ALKALINE HYDROLYSIS PROCESS FOR TREATMENT AND DISPOSAL OF PUREX SOLVENT ...

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ALKALINE HYDROLYSIS PROCESS FOR TREATMENT AND
DISPOSAL OF PUREX SOLVENT WASTE
by
C. Srinivas, K. A. Venkatesh, P. K. Wattal and T. K. Theyyunni
Process Engineering and Systems Development Division
and
P. K. S. Kartba and S. C. Tripathi
Fuel Reprocessing Division
1994
BARC/1994/E/019
ON
o
GOVERNMENT OF INDIA
ATOMIC ENERGY COMMISSION
ON
ON
ALKALINE HYDROLYSIS PROCESS FOR TREATMENT AND
DISPOSAL OF PUREX SOLVENT WASTE
by
C. Srinivas, K.A. Venkatesh, P.K. Wattal and T.K. Theyyunni
Process Engineering & Systems Development Division
and
P.i'.S. Kartha and S.C. Tripathi
Fuel Reprocessing Division
OHADHA ATOMIC RESEARCH CENTRE
liOMUAY, INDIA
BARC/1994/E/019
BIBLIOGRAPHY DESCRIPTION SHEET FOR TECHNICAL REPORT
(as per IS s 9400 - 19B0)
01 Security classification t Unclassified
02 Distribution : External
03 Report status : New
04 Series : BARC External
05 Report type : Technical Report
06 Report No. : BARC/1994/E/019
07 Part No. or Volume No. t
03 Contract No. :
10 Title and subtitle : Alkaline hydrolysis process -For
treatment and disposal of Purex solvent
waste
11 Collation : 31 p., 10 tabs., 1 fig.
13 Project No. :
20 Personal author (s) : 1) C. Srinivas; K.A. Venkatesh;
P.K. Wattalj T.K. Theyyunnif
2) P.K.S. Kartha* S.C. Tripathi
21 Affiliation of author (s) : 1) Process Engineering and Systems
Development Division, Bhabha Atomic
Research Centre, Bombay; 2) Fuel
Reprocessing Division, Bhabha Atomic
Research Centre, Bombay
22 Corporate author(s) : Bhabha Atomic Research Centre,
Bombay-400 085
23 Originating unit : Process Engineering and Systems
Development Division, BARC, Bombay
24 Sponsor(s) Name t Department of Atomic Energy
Type : Government
30 Date of submission : June 1994
31 Publication/ISSUB dati July 1994
contd...(ii)
(ii)
40 Publisher/Distributor i Head, Library and Information Division,
Bhabha Atomic Research Centra, Bombay
42 Form of distribution i
SB Language of text s
51 Language of summary s
52 No. of references :
53 Gives data en :
613 Abstract : Treatment of spent purex solvent (30%TBP-70% n-dodecane
mixture) from reprocessing plants by alkaline hydrolysis process was
investigated using inactive 30% TBP solvent as well as actual
radioactive spent solvent. Complete conversion of TBP to
water-soluble reaction products was achieved in 7 hours reaction time
at 130*C using 50Z<w/v> NaOH solution at NaOH to TBP mole ratio of
3:2. Addition of water to the product mixture resulted in the
complete separation of diluent containing less than 2 and 8 Bg./ml.
of a and (i activity respectively. Silica gel and alumina were found
effective for purification of the separated diluent. Aqueous phase
containing most of the original radioactivity was found compatible
with cement matrix for further conditioning and disposal.
70 Keywards/Dsscriptors : PUREX PROCESS} TBP; HYDROLYSIS; DODECANE;
RADIOACTIVE WAGIE PROCESSING; SODIUM HYDROXIDES; RADIOACTIVE WASTE
DISPOSAL; RADIDASSAYj SOLVENTS; SOLIDIFICATION; PURIFICATION;
PLUTONIUM; AIIERXCXUM* CEMENTS; RUTHENIUM 106; RADIOACTIVE EFFLUENTS;
SODIUM PHOSPHATES; DUTANOLS; LIQUID WASTES; SILICA 6EL; ALUMINIUM
OXIDES; AQUEOUS SOLUTIONS
71 Class No. : INIB Subject Category i E5100; ES200
99 Supplementary elements :
ABSTRACT
Treatment of spent solvent (Tributyl phosphate-dodecane
mixture) from reprocessing plants by alkaline hydrolysis process
was investigated.
Hydrolysis reaction was carried out by mixing the solvent
with sodium hydroxide solution of 50*(W/V) concentration at NaOH
to TBP mole ratio of 3:2. During hydrolysis, temperature of the
reaction mixture was maintained between 125 to 130*C under
constant stirring and total re.flux conditions.
Complete conversion of TBP( 99.99%) was achieved in about 7
hours of reaction time resulting in the formation of sodium
dibutylphosphate (NaDBP), sodium monobutyl phosphate (Na2MBP),
trisodium phosphate<Na3FO4)and butanol. NaDBP and butanol were
the principal reaction products.
Three layers were observed in the product mixture at the end
of the reaction. The top layer consisted of the diluent (n-
dodecane). The middle layer was an emulsion consisting of
unseparated diluent and the products of hydrolysis, viz., NaDBP,
Na2pQ4' Na3FO4 and butanol and also the unreacted NaOH. The
bottom thin layer was concenti'ated aqueous solution of unreacted
sodium hydroxide and hydrolysis products other than
butanol. 10- 40%(v/v) of the diluent originally present in the
solvent could be accounted for in the middle layer. About 90% of
the activity originally present in the spent solvent was
accounted for in -the bottom layer and the remaining 10% in thm
middle layer. The diluent layer was practically free of activity.
Addition of water to the reaction products followed by mixing
and settling resulted in the formation of only two layers and
complete separation of diluent. The top diluent layer contained
less than 100 ppm of TBP and was practically free from DBP, MBP
and butanol. The bottom aqueous layer consisted of the products
of hydrolysis of TBP and did not contain free TBP and DBP.
Several experiments were conducted using inactive TBF-
dodecane mixtures upto ten litre scale and actual radioactive
solvent waste upto two litre scale. Hydrolysis behaviour was
found to be identical and reproducible in all the cases ( active
and inactive).
In experiments with actual waste samples, the aqueous phase
retained practically all the radioactivity originally present in
the spent solvent. This phase was found to be compatible with
cement inatri:-: for conditioning and disposal.
The diluent product contained less than 2 Bq./ml. of gross
alpha and less than 8 Bq./ml. of gross beta activity. Further
purification with silica gel / alumina reduced the radioactivity
in the diluent by a factor of 3 to 8. The Plutonium retention
value of the purified diluent was comparable to that of fresh
diluent. Thus the diluent so separated and purified can be
recycled and reused.
ALKALINE HYDROLYSIS PROCESS FOB TREATMENT AND DISPOSAL 07
FOREX SOLVENT WASTE
by
C. Srinivas, K.A. Venkatesh, P.K. Wattal, T.K. Theyyunni
and
P.K.S. Kartha and S.C. Tripathi
1 IntroductIon
In the Purex process, a mixture of tri-n-butyl
phosphate(TBP) and a diluent (Shell Sol T, a mixture of C-12 to
C-18 saturated hydrocarbons, or n-Dodecane) is used as solvent
for extraction and purification of uranium and Plutonium from
spent fuels. During its use over a period of time, the solvent
undergoes chemical and radiolytic degradation which adversely
affect the extraction process and hence is discarded as a waste
[1-4].
Various processes have been proposed for treatment and
disposal of this organic solvent waste. Direct immobilization in
organic/cement matrices as a mode of conditioning and disposal
has been reported [5,6]. Direct incineration has also been
attempted as a treatment mode. Use of H3PO4 has been tried for
extraction of TBP by adduct formation and separation of diluent
[7,8]. Destruction of TBP by wet oxidation using H2O2 was also
investigated in our laboratory [9]. Complete oxidation of TBP to
COg, water and H3PO4 at 100°C and atmospheric pressure was
achieved in the presence of iron salt as catalyst. Separated
diluent contained only 10% of the original radioactivity. Wet
oxidation process has certain attractive features such as mild
operating conditions, simple treatment of off-gases and
corrosion-free conditions.
However, generation of large volume of secondary aqueous
waste is its major disadvantage.
Alkaline hydrolysis reaction of TBP is well known[10].
Kinetic and mechanistic aspects of this reaction have received
considerable academic attention. The process is simple for
adoption in plant scale. It has a better economic viability
compared to other processes. Only small volumes of secondary
wastes are generated.
D-5t.ai.lc5d investigation of the alkaline hydrolysis process
for treat-Men t of the pur<sx solvent waste was taken up on
laboratory and bsiidi scales to assess and evaluate its
suitability for adoption in the plant scale. Results of these
investigations are presented in this report.
2 Kxperiiaantal •nrooediire
Experiments were conducted using inactive solvent upto 10
litre scale and actual spent solvent samples upto two litre
scale. The radiochemical composition of two typical actual
solvent waste samples used in experiments is iven in Table-1.
Procedure for alkaline hydrolysis followed is given below:
Known volume of tfc<3 solvent sample to be hydrolysed was mixed
with 50XO//V) aqueous sodium hydroxide solution keeping mole
ratio of U&OiUTBP::3:2 in a glass flask provided with reflux
condenser and stirred at a constant speed. For 200 ml. samples,
stirring was done magnetically. Mechanical stirrer was used for
experiments with more than 200 ml. samples. Temperature of the
reactant mixture was maintained between 125 to 130*G for 7 hours
2
under total reflux conditions [11]. The reaction flask was
immersed in a hot paraffin oil bath to maintain the desired
reaction temperature.
After completion of reaction time, the contents of the
reaction flask were allowed to cool to room temperature. Water
was added to the product mixture, stirred for 15-20 minutes to
completely separate the diluent from the highly emulsified
aqueous phase. The separated diluent layer was analysed for the
presence of TBP and DBP. Aqueous phase was analysed for the
products of hydrolysis, viz., sodium dibutyl phosphate(NaDBP),
sodium monobutyl phospii&te(Wa2MBP), tri sodium phosphate (Na3PO4>
and butanol as well as for TBP and DBP. In experiments with
actual radioactive solvent waste samples, radiochemical analyses
were also carried out to determine gross alpha and beta
activities present in the diluent. Fig.1 shows the
experimental set-up?used for hydrolysis studies.
3 Purification oj£ the product diluent.
To explore the possibility of recycling the diluent product
for reuse in reprocessing plants, the diluent was subjected to
a series of purification steps as follows:
It was contacted with sorbents such as silica gel, alumina,
macroreticular anion-exchange resin, Amberlyst-26( in 0H~ form )
and hydrous titania. In some experiments, before contacting
with the sorbents, the diluent was first given a series of washes
with alkaline potassium permanganate,dilute nitric acid and
water. Both column and batch modes were used for purification of
the diluent with solid sorbents. In batch mode, 5.0 ml. of
diluent was mixed with 0.25 grams of the sorbent and shaken for
30 minutes. The diluent was then separated by filtration. In
column mode, about 50-70 bed volumes of diluent were passed
through the sorbent at the rate of 0.25 to 0.5 bed volumes per
minute. Experimental details for column mode purification were
shown in Table-2. These purified diluents were subjected to
Plutonium retention test (sect.4).
4.Plutonium retention test for the product diluent
Plutonium retention test was conducted to determine the
purity of the diluent with respect to products of degradation
that might affect its performance during-solvent extraction. A
Plutonium stock solution of about 0.4 mg. per ml. was prepared in
3M HNO3 and was conditioned to Pu(IV) state. 100 uL of this
stock solution was added to 3.0 ml.of 3M HNO3 and this mixture
was shaken with equal volume of diluent product for 20 minutes.
After phase separation, Plutonium in both the layers was analysed
by direct planchetting and alpha counting. Diluent layer was
then stripped by shaking for 10 minutes with 0.1M .pa nitric acid
in 1:2 volume ratio. Four such stripping contacts were
given. Plutonium in the diluent phase after these strippings was
estimated by direct planchetting. Percentage Plutonium retention
in the diluent was calculated as follows:
Pu in diluent
after stripping
% Pu retention = x 100
Sum of Pu contents in diluent and
aqueous phases before stripping